[en] Simulations of the breakup and penetration of high temperature thermite jets into water have been performed using the integrated fuel-coolant interaction code, IFCI. The work to date has been directed towards assessing the model performance against data obtained from an experimental test series performed at Sandia. The tests, part of the EJET series, were extensively photographed to allow for direct digitization of the melt profile data thereby allowing a direct comparison of the IFCI predictions to the test data. This document is a preliminary report for Task 1, Molten Jet Model Evaluation, of the Molten Fuel-Coolant Interaction Program. The IFCI simulation of test EJET-1, with initially saturated water, showed reasonable performance in predicting early time leading edge penetration rate and initial jet spreading as shown by comparisons of the molten thermite volume fraction. A transition to a bulk boiling temperature regime which was observed in both tests was not modeled adequately by IFCI. An attempt to simulate test EJET-0, with initially subcooled water, failed at very early times due to an automatic decrease in the time step to an unacceptable value caused by nonconvergence of the numerical algorithm. The preliminary assessment results suggest (1) the need to include the steam volume fraction in the data comparisons, (2) an improvement of the boiling model in IFCI to address the bulk boiling question, and (3) use of a finer noding scheme to improve the spatial resolution of IFCI before the code is applied to addressing accident management concerns at reactor scale

[en] Probabilistic risk assessment (PRA) studies are being extended to include a wider spectrum of reactor plants than was considered in NUREG-1150. There is a need for simple direct containment heating (DCH) models that can be used for screening studies aimed at identifying potentially significant contributors to overall risk in individual nuclear power plants. This paper presents two adiabatic equilibrium models suitable for the task. The first, a single-cell model, places a true upper bound on DCH loads. This upper bound, however, often far exceeds reasonable expectations of containment loads based on CONTAIN calculations and experiment observations. In this paper, a two cell model is developed that captures the major mitigating feature of containment compartmentalization, thus providing more reasonable estimates of the containment load. Predictions of the equilibrium models are compared with experimental data from the Limited Flight Path test series conducted at Sandia National Laboratories

[en] Institute for Nuclear Protection and Safety (IPSN) has developed a rationale for accident studies which involves both analytical and crisis strategies. The operational aim to provide as high as possible prevention of damage for installations and environment is fulfilled during accidental and post-accidental phases through development of crisis tools and analysis of emergency plans. Further research will provide still more detailed insight into release prevention capabilities and environment recovery techniques

[en] The need for long term development of accident management programs is acknowledged and the key tool for that development is identified as the IPE Program. The Edison commitment to build an integrated program is cited and the effect on the IPE effort is considered. Edison's integrated program is discussed in detail. The key benefits, realism and long term savings, are discussed. Some of the highly visible products such as neural network artificial intelligence systems are cited

[en] The spreading, cooling, and freezing of molten core debris on a horizontal surface during a postulated severe accident are important considerations which influence the containment thermal response. This study describes theoretical models for predicting the time-dependent spreading geometry of molten core debris on a horizontal floor, and several associated cooling responses. It was found that corium discharge from a doorway tends to have a spreading angle of about 52 degrees when surface tension is negligible. Simplified heat transfer models are employed to estimate local freezing and mounding of flowing corium, which can diminish its coolability. Effects of metal/oxide stratification and voids on the hot spot temperature also are included. It is shown that when corium arrives at a wall, the resulting hot spot temperature is reduced if the wall slopes away from the corium

[en] An overview on Japanese activities of severe accident research is presented, covering various fields and topics of experimental investigation on severe accident phenomena such as fuel damage and melt progression, fission products release and transport, and component and containment integrity. The current status of analytical investigation on severe accident is also described in the fields of the level-1 and level-2 probabilistic safety assessment (PSA) studies, code development and assessment activities. The basic considerations on accident management are summarized

[en] The purpose of this paper is to discuss the US nuclear industry activities, occurring under the auspices of Nuclear Management and Resources Council (NUMARC), to define, develop and implement enhancements to utility accident management capabilities. This effort consists of three major parts: (1) Development of a practical framework for evaluation of plant-specific accident management capabilities and the subsequent implementation of selected enhancements. (2) Development of specific technical guidance that address arresting core damage if it begins, either in-vessel or ex-vessel, and maintaining containment integrity. Preventing inadequate core cooling or minimizing the consequences of offsite releases, while considered to be candidate areas for accident management enhancements, have been the subject of intense previous study and development. (3) Plant-specific implementation of accident management enhancements in three areas: (a) personnel resources (organization, training, communications); (b) systems and equipment (restoration and repair, instrumentation, use of alternatives); and (c) information resources (procedures and guidance, technical information, process information)

[en] This three-volume report contains 83 papers out of the 108 that were presented at the Nineteenth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 28--30, 1991. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 14 different papers presented by researchers from Canada, Germany, France, Japan, Sweden, Taiwan, and USSR. This document, Volume 2, presents papers on: Severe accident research; Severe accident and policy implementation; and Accident management. The individual papers have been cataloged separately

[en] Although MELCOR is now being successfully applied in severe accident analyses, it is not yet complete and additional development and assessment is needed before MELCOR can fully satisfy its design objectives and be applied with confidence to its targeted applications. A number of current and planned improvements and assessment activities necessary to reach that stage are described in this paper. Modifications that have been implemented in the latest release of the code, version 1.8.1, are summarized, the status of work in progress on new models such as direct containment heating, in-vessel natural circulation, and materials interactions is given, and several additional models and other enhancements planned for the near future are described. The results of recent assessment calculations performed at Sandia National Laboratories are summarized, and assessment efforts that have just begun or are planned for the near future are briefly mentioned

[en] Two revisions of the CONTAIN code, CONTAIN 1.11 and 1.12, have recently been released. The purpose of this paper is to highlight the new features of these revisions and to discuss other new code features currently under development. The features of CONTAIN 1.11 discussed here include a quasi-mechanistic concrete outgassing model, the connected structure option for heat conduction between compartments, and a new approach for modeling forced convective heat transfer. The direct containment heating (DCH) models released as part of CONTAIN 1.12 are also discussed. New code features currently under development include a revised gas combustion model and a new multifield DCH model. New features of the revised combustion model include the treatment of spontaneous recombination and diffusion flames. CONTAIN plant calculations comparing the old and the revised combustion models are presented. The new features of the multifield DCH model are discussed, and demonstration calculations using this model to analyze a small scale experiment are presented